This report deals with a CORA-9 post-test-calculation with the computer-code RELAP5/SCDAP Mod.2.5 Ver.3f. It is the second technical report concerning the BMFT founded project ''Comparative Assessment of Different Computer Codes for Severe Accident Analys...

... Accession Number : AD0620258. Title : FURTHER ATTEMPTS AT CODING AIRCRAFT ACCIDENTS. Descriptive Note : Research rept.,. ...

... AD0076065. Title : CODED FLEET ACCIDENTS OF RECENT GRADUATES. Corporate Author : TULANE UNIV NEW ORLEANS LA. ...

It is prerequisite for the establishment of nuclear power plant safety and for the maximization of operation efficiency to devlope the accident analysis computer code packages which can predict the results of postulated accidents and evaluate the performa...

The VICTORIA code is designed to be a module of the MELPROG severe accident analysis code. The VICTORIA code is designed to model the in-vessel phase of a severe reactor accident. The functions of VICTORIA in MELPROG are fission product release from fuel,...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was developed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was developed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

The MARCH code, written at Battelle's Columbus Laboratories for the U.S. Nuclear Regulatory Commission, describes the response of LWR systems to accidents which can result in core meltdown. The calculations are performed from the start of the accident thr...

An exploratory sensitivity study with the MELCOR Accident Consequence Code System (MACCS) is presented. This study was performed to provide (1) an indication of the possible impact of consequence modeling uncertainties on the results of an integrated prob...

In the analysis of accidents a collision between aircraft is treated as one accident in the overall total. However, a complete analysis and coding is made on each aircraft involved in a collision. This produces two aircraft accident records, one for each ...

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system and the core during a severe accident transient, ...

A computer code called TRAC is being developed by the Los Alamos Scientific Laboratory for analysis of loss-of-coolant accidents (LOCA's) and other transients in light water reactors. This code differs from existing codes and other codes under development...

... Title : Uncertainty and Sensitivity Analysis of a Fire-Induced Accident Scenario Involving Binary Variables and Mechanistic Codes. ...

This paper is concerned with the application of severe-accident codes to develop improved accident mitigation strategies for nuclear power plants employing boiling water reactors (BWRs). Specifically, the results of analyses using the APRIL.MOD3X code are presented for typical station blackout ...

It is not feasible to use the existing British Coal accident information collection, storage and retrieval system to identify systematically human factors issues relevant to the accident. The implementation of a human factors code list, developed during t...

A key parameter in the calculation of accident dose-risks by the RADTRAN 4 code is the time assigned for evacuation of the affected area surrounding the accident. Currently, in the interest of assured conservatism, this time is set at 24 hrs. Casual anecd...

This report describes the analytical results of operational transient and accidents for the safety assessment of upgraded JRR-3. The following six items of transients and accidents have been selected and analyzed for the assessment; 1. decrease of primary...

The MARCH and HECTR computer codes are used in this study to examine hydrogen production, transport, and combustion in an ice-condenser containment for a number of hypothesized severe accidents. Both degraded-core and core-meltdown accidents are treated. ...

The MELCOR Accident Consequence Code System (MACCS) has been applied to the radiological consequence assessment of potential accidents from a non-reactor nuclear facility. MACCS has been used in a variety of applications to evaluate radiological dose and ...

The Department of Energy has developed new capabilities to predict how an accident event would spread through a facility. These capabilities arose as a part of the development of a future generation accident code - CONACS. They can be used to predict the ...

After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into ...

One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat t...

Cheap and clean energy can be supplied with nuclear power plant whose accident probability is very low comparing with that of the other industrial facilities. However, the consequences of severe accident of nuclear power plant may result in a critical imp...

This report provides documentation for UMTRIs file of Buses Involved in Fatal Accidents (BIFA), 2008, including distributions of the code values for each variable in the file. The 2008 BIFA file is a census of all buses involved in a fatal accident in the...

This report provides documentation for UMTRIs file of Buses Involved in Fatal Accidents (BIFA), 2007, including distributions of the code values for each variable in the file. The 2007 BIFA file is a census of all buses involved in a fatal accident in the...

The publication, containing reports of U. S. Civil aircraft accidents, in brief format, is issued monthly and includes those accidents analyzed and processed during the preceding month. The reports are reproduced directly from the coded record on magnetic...

The publication, contains reports of U. S. Civil aricraft accidents, in brief format, is issued monthly and includes those accidents analyzed and processed during the preceding month. The reports are reproduced directly from the coded record on magnetic t...

A computer modeling code, CRIT8, was written to allow prediction of the radiological doses to workers and members of the public resulting from these postulated maximum-effect accidents. The code accounts for the relationships of the initial parent radionuclide inventory at the time of the accident to the growth of ...

In this paper, a simple analytical code, QUASAR, is developed to analyze the phenomena related to severe subassembly accidents, such as a total instantaneous blockage event for a subassembly inlet. The code models failed and neighboring subassemblies, focusing mainly on the thermal consequences and the propagation potential of the ...

The MELT-III code for analysis of a Transient Overpower (TOP) accident in an LMFBR is briefly described, including failure criteria currently applied in the code. Preliminary results of calculations exploring failure patterns in time and space in the reac...

The SAS4A code system is a new tool for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failures of the subassembly walls. The objective in the development of SAS4A is to provide improved analytical mod...

During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident an...

The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accid...

Design-basis depressurization accident analyses for the Fort St. Vrain reactor were performed using the FLODIS (Ref. 4) code. The FLODIS code models the active core, side reflector, gas annulus between the core barrel and the PCRV liner, and the PCRV cool...

This paper summarizes a recently completed study that identified and investigated the difficulties and limitations of applying first-order adjoint sensitivity methods to reactor accident codes. The work extends earlier adjoint sensitivity formulations and...

The accident at Three Mile Island Unit 2 (M-2) provides an opportunity to benchmark severe accident analysis methods against full-scale, integrated facility data. In collaboration with the US Department of Energy (DOE), the OECD Nuclear Energy Agency established a joint task group to analyze various periods of the accident and ...

The accident at Three Mile Island Unit 2 (M-2) provides an opportunity to benchmark severe accident analysis methods against full-scale, integrated facility data. In collaboration with the US Department of Energy (DOE), the OECD Nuclear Energy Agency established a joint task group to analyze various periods of the accident and ...

Two computer codes are developed for performing the calculations of SUPER-PHENIX integrity when subject to an hypothetical core disruptive accident; both codes are 2D axisymmetric but they use different meshes: Lagrangian for the SIRIUS code, Eulerian and...

Sandia has made considerable progress in the past year on the MELCOR code for integrated severe nuclear reactor accident analysis. Actinities for the past year are presented.

A hazard assessment was conducted on the GRC managed payloads in support of a NASA Headquarters Code

... ACCIDENTS, UNITED STATES, INTEGRATED SYSTEMS, INDUSTRIES ... CASUALTIES, PERSONNEL, WORK, MORTALITY RATE, AGRICULTURE. ...

SECTION 1 - AMICABLE ADJUSTMENT OF DIFFERENCES: In the belief that all ...... may be granted by supervisors for illness, marriage, accident, or critical ...

This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in ...

The capability of the RELAP5/MOD3 code to validate various transients encountered in RBMK reactor postulated accidents has been assessed. The assessment results include a loss of coolant accident at the inlet of the core pressure tube, the blockage of a pressure tube, and the pressure response of the core cavity to in core pressure ...

Fission product release analysis code, RACPAC (Fission Product Release Analysis Code from Fuel Particle in Accident Condition), was developed to calculate fractional release from the core during accident conditions of High Temperature Gas-cooled Reactor. ...

The ORION code is designed to determine very quickly the immediate consequences (such as plume passage time, instantaneous maximum hazards irradiation, inhalation, deposit) due to an accident spreading out radioactive or chemical pollution into the atmosp...

The purpose of this paper is to present and document the differences discovered when comparing the two accident dose codes FUSEDOSE and MACCS2. Each code`s methodology is first discussed. With this background, the important comparison parameters are discussed and the resulting differences are presented. It is not the purpose of this ...

This paper describes methods, in the form of a handbook and five computer codes, that can be used for air cleaning system design and accident analysis. Four of the codes were developed primarily at the Los Alamos National Laboratory, and one was developed in France. Tools such as these are used to design ventilation systems in the ...

The implementation of SEDA 2.0 computer code, developed at Ezeiza Atomic Center, Argentine for Angra 1 reactor is described. The SEDA code gives an estimate for radiological consequences of nuclear accidents with release of radiactive materials for the en...

The TRAC (Transient Reactor Analysis Code) program is an advanced computer code which is designed to model postulated accidents in light water reactors. The paper discusses some of the methodology employed in modeling large internal flow systems. Emphasis is placed on the numerical and modeling problems inherent in computer ...

A number of computer codes have been written to attempt to predict the consequences of hypothetical accidents. These codes all have one thing in common: they produce reams of paper containing vast quantities of numbers. In an effort to make these numbers ...

CONTAIN is a computer code to analyze physical, chemical and radiological processes inside the reactor containment in the sequence of severe reactor accident. Modelling of the aerosol behavior is included. We have improved the code by implementing a subro...

Several computer codes have been developed to predict the behavior of aerosols in steam, a situation which is expected to occur in a light-water reactor accident. Among the codes with capabilities in this respect are MAEROS (Sandia) AEROMECH (University o...

The VICTORIA code is designed to be a module of the MELPROG severe accident analysis code. The functions of VICTORIA in MELPROG are fission product release from fuel, vapor-phase and aerosol physics, and chemical interactions involving fuel, structures, c...

The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the ...

APRIL is a fast-running and user-friendly system code for interactive simulations of severe accidents in boiling water reactors (BWRs). The component models in the most recent version, APRIL.MOD3X, include the reactor core and pressure vessel, as well as the primary and secondary containments. Whereas APRIL.MOD3X is a fast-running ...

The purpose of the work described in this paper was to evaluate the adequacy of criticality accident alarm system (CAAS) coverage of several buildings at the Portsmouth Gaseous Diffusion Plant (PORTS) located in Piketon, Ohio. An analysis was performed in which the General Monte Carlo N-Particle Transport Code (MCNP) was used to model several of the ...

The RELAP5/MOD1 computer code was used in the analysis of a loss coolant accident (LOCA), postulated for Angra-2. The power plant was simulated through a division in control volumes and junctions suitable for a small break accident calculation. Initially ...

Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes used in impact analysis. (LN)

The paper discusses Nuclear Regulatory Commission (NRC) efforts regarding severe reactor accident management and the Nuclear Management and Resources Council (NUMAEX), activities. (EPRI) Electric Power Research Institute accident management program consists of the two products just mentioned plus one related to severe accident plant ...

. Although analyses of HFACS-coded data reveal that CRM is a factor in aviation accidents, the actual1 Paper Number 2005-xx-yyyy CLASSIFYING CREW PERFORMANCE FAILURES IN COMMERCIAL AVIATION ACCIDENTS University of Illinois at Urbana-Champaign Champaign, IL The purpose of accident investigation is to identify

INTRODUCTION: The rate for work related accidents in the Spanish mining sector is notably higher than in other countries such as the United States. It produces a very negative impact on the mining industry. This paper is the report of a study on serious and fatal accidents in Spanish mining from 1982-2006. It is based on the reports of 212 ...

The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the ...

A FORTRAN computer program, HAZARD, was written and programmed for the CDC 6600 to calculate infinity does to the thyroid resulting from exposure to a reactor plume following a design basis accident. Using, as input, plant operating characteristics and me...

The FORTRAN-IV computer code UFOMOD calculates the radiological consequences of reactor accidents for risk studies, namely early deaths, latent cancer deaths and genetic significant doses. Different models for the atmospheric transport and deposition, the...

This report provides documentation for UMTRIs file of Trucks Involved in Fatal Accidents (TIFA), 2006, including distributions of the code values for each variable in the file. The 2006 TIFA file is a census of all medium and heavy trucks involved in a fa...

The rod ejection accident (REA) of the Beznau (KKB-2) nuclear power plant was investigated. The REA analysis was performed using the RETRAN-02 computer code. Four basic cases were investigated for the cycle 16 conditions. At the beginning-of-life (BOL) th...

Sensitivity analyses were performed in order to investigate the effect of reflooding on the hydrogen generation and temperature rise of the fuel cladding due to Zr-water reaction. The sensitivity analyses were conducted for the TMI-2 accident assuming tha...

This report documents the results of an assessment to determine the reactor and containment hardware, systems, and phenomena which must be modeled in realistic boiling water reactor severe accident analysis computer codes. The scope of the assessment is l...

This study is an evaluation of a pressurized-water reactor (PWR) accident as defined by WASH 1400 for the proposed nuclear reactor site at Cementon, N. Y. Using an extension of the Environmental Protection Agency's AIREM computer code, the following were ...

The joint USNRC/CEC consequence uncertainty study was chartered after the development of two new probabilistic accident consequence codes, MACCS in the U.S. and COSYMA in Europe. Both the USNRC and CEC had a vested interest in expanding the knowledge base...

This volume is the second of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. Thes...

This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These...

The Three Mile Island nuclear plant (TMI-2) was modeled using the Transient Reactor Analysis Code (TRAC-PIA) and a preliminary calculation, which simulated the initial part of the accident that occurred on March 28, 1979, was performed. The purpose of thi...

and grading code. This is useful as a basis for decision-making in the medical management as it assigns of the severity of damage � the decision on the kind of hospital � the provision of appropriate therapeutic therapeutic procedures it is necessary to assess the state and probable outcome of radiation accident victims

The MELPROG computer code is being developed to provide mechanistic treatment of Light Water Reactor (LWR) accidents from accident initiation through vessel failure. This paper describes a two-dimensional (r-z) debris meltdown model that is being develope...

Sandia National Laboratories is participating in several NRC-sponsored programs to study severe accident phenomenology. Part of that effort involves the combined use of the computer codes MARCH and HECTR to examine hydrogen behavior during severe accident...

MACCS2 represents a major enhancement of the capabilities of its predecessor MACCS, the MELCOR Accident Consequence Code System. MACCS was developed to estimate the potential impacts to the surrounding public of severe accidents at nuclear power plants. T...

The report documents the procedures and results of an analysis of preventable accidents involving motor carriers. The process used to assign 'fault' to drivers, vehicles, and owners is explained. Also, data sources, coding conventions and analytical techn...

Although hypothetical fast reactor accidents leading to severe core damage are very low probability events, their consequences are to be assessed. During such accidents, one can envisage the ejection of sodium, mixed with fuel and fission products, from t...

The CRAC2 computer code has been adapted to the calculation requirements of Draft/Final Environmental Impact Statement (DES/FES) casework analysis for the Nuclear Regulatory Commission. CRAC2 is a revised version of the CRAC (Calculation of Reactor Accide...

CORCON is a computer code for modelling the interactions between molten core materials and concrete, such as might occur following a core meltdown accident in a Light Water Reactor. It may also be applied to experiments which simulate such accident condit...

CONTAIN is an integrated containment thermohydraulic and aerosol transport code. The choice of the Surry plant provides an opportunity to compare the CONTAIN calculations to those recently performed by Battelle Columbus Laboratories. The accident sequence...

The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. Calculations of this design-basis event has been done conservatively because there was margin to the fuel failure criterion of 17...

Aerosol behaviour in the reactor containment was studied in the case of severe reactor accidents. The study was performed in a Nordic group during the years 1985 to 1988. Computer codes with different aerosol models were used for calculation of fission pr...

A thermal-hydraulic analysis to describe the transient behavior in the water loop of the Nuclear Engineering Department of the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brazil, was performed. Postulated accidents such as those resulting f...

This paper describes a multilaboratory research program that is directed toward addressing many questions that analysts face when performing air cleaning accident consequence assessments. The program involves developing analytical tools and supportive experimental data that will be useful in making more realistic assessments of accident source terms within ...

The RELAP5/MOD3 computer code has been developed for best-estimate simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents...

have been examined using the codes RELAP, TRAC, RAMONA, and SMABRB. Severe accidents have been-of-coolant accident analysis. Pour differ- ent codes have been studied during 1982. The American code RELAP5 has been and to the RELAP5 calcu- lations of Finland and Sweden. The comparison to data ...

THALES-CV1, a computer code for evaluating reactor containment pressure and temperature during core meltdown accident, was developed. In the code a whole free volume inside the containment is divided into several number of compartments which are connected...

Cloning by Accident: An Empirical Study of Source Code Cloning Across Software Systems Raihan Al that pertain to a software system and its problem domain. Source code cloning is one way in which expertise can be reused across systems; cloning is known to have been used in several open source projects

This document has been prepared as a user's guide for the computer program RISKIN developed at Sandia National Laboratories. The RISKIN code generates integrated risk tables and the weighted mean risk associated with a user-selected set of consequences from up to five output files generated by the MELCOR Accident Consequence ...

The computer code CITADEL was written to analyze iodine behavior during steam generator tube rupture accidents. The code models the transport and deposition of iodine from its point of escape at the steam generator primary break until its release to the environment. This report provides a brief description of the ...

For many years, study of whole-core transients and accidents in fast reactors has been a major element of safety research and development programs in the US. Numerous models and computer codes have been developed and validated for use in these studies. Historically, emphasis has been placed on describing the core disruptive accident, ...

This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 ...

A hypothetical accident is analyzed, in which an external (out-of-plant) natural or man-made event causes a loss-of-coolant accident after penetrating the containment wall. The computer codes CONTEMPT and RELAP4 have been used to study the containment thermal-hydraulic behavior during the accident. Results are ...

This study evaluates the effect that the timing of vessel depressurization has in the progression of a short-term station blackout accident in a boiling water reactor (BWR)-4. This study was performed with the MELCOR (version 1.8.1) severe-accident code. A similar study was previously completed at Oak Ridge National Laboratory (ORNL) ...

... was heightened by the accident at the USA's Three Mile Island plant. ... for ASME XI Appendix VIII the US code covering performance demonstration. ...

... does to the thyroid resulting from exposure to a reactor plume following a design basis accident. Using, as input, plant operating characteristics and ...

... 1 had an incidence of 32.5%, compared to 41.8% for ... Ethnicity was coded as white vs. ... Smoking, driver education, and other correlates of accidents ...

MARCH 2 calculations for the TMLB' accident (a transient with loss of all ac power) at Surry provide a test bed for addressing problems with computer models in an analysis situation typical of those expected in applications of the MELCOR code system for severe accident analysis. This paper describes a sensitivity analysis for ...

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the saf...

codes Q and M. The Challenger accident. (1986) impacted the development of this technology in several ways. First, Professor. Richard. Feynman, ...

This report provides background information for a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium blanket heat removal systems.

The facilities to influence the interaction between melt and the concrete foundation by a sacrificial layer was investigated by the help of the code SACRI. The results of this investigation show that a relative moderation of the course of the accident can...

is analyzing response of paratranist bus structure to rollover accident using LS-DYNA code. Click for Details Transims visualization Transims visualization TRANSIMS...

... The computer codes RELAP-3 for reactor system blowdown and THETAl-B for fuel element heatup were used to perform these calculations at a ...

May 1, 2011... generates reports that are made available to the public at no cost. ... the Presidential Commission on the Space Shuttle Challenger Accident. ...

... care benefits in the event of accident or sickness. SB 1237 Brings pharmacy benefit managers under Texas Insurance Code Article 21.07-6, ...

Regardless low probability of occurrence the severe accident phenomena are investigated for all types of nuclear reactors in the world because the consequences of such accident could be catastrophic. Most of research is performed for the prevailing vessel-type light water reactors like PWRs and BWRs. Less research is performed for the channel-type reactors ...

To ensure that the public health and safety are protected under any accident conditions in a Liquid Metal Fast Breeder Reactor (LMFBR), many accidents are analyzed for their potential consequences. The SAS4A code system, described in this paper, provides such an analysis capability, including the ability to analyze low probability ...

The MELCOR code, developed by Sandia National Laboratories, is a fully integrated, relatively fast-running code that models the progression of severe accidents in commercial light water nuclear power plants (NPPs).A specific station blackout (SBO) accident for Kuosheng (BWR-6) NPP is simulated using the MELCOR ...

An assessment of the radiological consequences that would result for the population of Greece from postulated major nuclear accidents in the Kozloduy nuclear power station in Bulgaria is performed. Kozloduy lies at a distance of 225 km from the northern borders of Greece and contains six reactors, all of the Russian WWER type. The postulated accidents that ...

The MELCOR severe accident code version 1.8.4QK has been employed to calculate the timing of events and source terms released into the environment for several hypothetical severe accidents as Unit 1 of the Juragua VVER 440-213 plant, located near Cienfuegos, Cuba. Different severe accidents were simulated, and ...

This report presents the comparisons between calculated and measured fuel rod behavior and the analysis of stress for preirradiated LWR type fuel rods during reactivity initiated accident (RIA) conditions. For the calculations, FPRETAIN computer code whic...

A three-dimensional reactor dynamics computer code HEXTRAN has been developed, thoroughly validated, and extensively applied for transient and accident analyses of VVER type nuclear reactors. HEXTRAN models accurately the VVER core with hexagonal fuel ass...

APRIL is a mechanistic core-wide meltdown and debris relocation computer code for Boiling Water Reactor (BWR) severe accident analyses. The capabilities of the code continue to be increased by the improvement of existing models. The report contains inform...

VICTORIA is a mechanistic computer code designed to analyze fission product behavior within a nuclear reactor coolant system (RCS) during a severe accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from t...

The report describes the technical bases and use of two updated versions of a computer code initially developed to serve as a tool for calculating aerosol particle retention in boiling water reactor (BWR) pressure suppression pools during severe accidents...

Accurate predictions for the normal, off-normal or accident conditions are required for the safety analysis to evaluate the plant safety. Safety analysis computer codes such as IANUS, DEMO have been specifically developed for the Fast Flux Test Facility, ...

An update is given on very recent developments involving CONTAIN, the USNRC's principal mechanistic code for severe accident containment analysis. First, the features are outlined in two major new releases of CONTAIN. Revision 1.06 was released in Februar...

The RELAP5 independent assessment project is part of an overall effort to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. The RELAP5/MOD1 code has been as...

The RADionuclide Transport, Removal, And Dose (RADTRAD) code is designed for US Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the offsite population and to control room operators following a design-basis accident ...

The RADTRAN 4 computer code is designed to analyze radiological consequences and accident risks of transporting radioactive material. This manual provides information useful for interpreting, troubleshooting, or debugging components of the code during development or revision of the program.

The RADTRAN 4 computer code is designed to analyze radiological consequences and accident risks of transporting radioactive material. This manual provides information useful for interpreting, troubleshooting, or debugging components of the code during development or revision of the program.

An analytical model is presented for the calculation of pressure and temperature variations inside an ice condenser containment during the depressurization and post-blowdown phase following a loss-of-coolant accident. The method applies conservation equat...

An analytical model is presented for the calculation of ice melt and long-term pressure transients in the ice condenser reactor containment after a loss-of-coolant accident. The method applies conservation equations to control volumes used to simulate spa...

This report describes the technical bases and use of computer code ICEDF. ICEDF was developed to serve as a tool for calculating particle retention in pressurized water reactor (PWR) ice compartments during severe accidents. This report also serves as a c...

This report describes development of a revised Savannah River Site (SRS) meteorological data set for the MELCOR Accident Consequence Code System (MACCS). This data set contains quality assured values of transport wind direction, wind speed, atmospheric st...

Technical progress during this quarter has been concerned primarily with continued development of the TRAP-MELT computer code, specification of conditions and code input for meltdown accidents, and initiation of efforts to provide functional design specif...

An IBM 704 code concerned with loss of flow accidents in a three loop cooling system is described. The statement of the physical problem, equations solved, method of solution, input preparation, and machine operation are includcd. (auth)

A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and ...

The characteristics the flow of water in the hydraulic path of the multipurpose valve with the parameters of coolant had at the moment when the accident began in Unit 4 at the Chernobyl nuclear power station are determined using a 3D thermohydraulic code.

Status of significant computer codes developed within the Project 'Core Meltdown' of the Light-Water Reactor Safety Research Program funded by the Federal Ministry of Research and Technology (BMFT). The Project Core Meltdown was initiated as basic researc...

The WECHSL code, developed at Kernforschungszentrum Karlsruhe, West-Germany, is used for core melt accidents in nuclear power plants. The first calculations, considering silicate and limestone/common sand concretes of different compositions, analyze the i...

The SIMMER Code, developed at Los Alamos Scientific Laboratory, analyzes the various phases of a CDA (Core disassembly accident) in an LMFBR. The qualification testing program, comprised of two stages, is discussed. The first stage is developmental verifi...

The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident compute...

Accident analyses are being performed to evaluate and document the safety of the Hanford Waste Vitrification Plant (HWVP). The safety of the HWVP is assessed by evaluating worst-case accident scenarios and determining the dose to offsite and onsite receptors. Air dispersion modeling is done with the GENII computer code. Three ...

This paper provides an overview of the current activities within the international consortium participating in the Three Mile Island Unit 2 (TMI-2) analysis exercise, which is part of the Nuclear Energy Agency/Department of Energy Joint Task Group (JTG) program on TMI-2, formed to utilize the TMI-2 accident as a benchmark for severe-accident computer ...

This paper concerns applicability of drywell cooler (DWC) heat removal under severe accident condition in BWR plants. Newly developed heat removal models based on DWC heat removal experiments were built into the MAAP3 code. And then, two types of Japanese BWR were selected to evaluate DWC heat removal performance under typical severe ...

An international analysis exercise has been organized to evaluate the ability of nuclear reactor severe accident computer codes to predict the TMI-2 accident sequence and core damage progression during the first 300 minutes of the accident. A required bou...

The Modular Accident Analysis Program (MAAP) simulates LWR system response to a severe core accident. Overall, calculations performed with the PWR version of MAAP have compared well with a wide variety of other data. These results have proven MAAP an acceptable tool to support individual plant examinations and accident management. This ...

The US Department of Energy's High Flux Isotope Reactor (HFIR) was analyzed to evaluate it's response to a spectrum of loss-of-coolant accidents (LOCAs) with potential for leading to core damage. The MELCOR severe accident analysis code (version 1.7.1) wa...

We have surveyed new technologies and research results for the accident management of nuclear power plants. And, based on the concept of using the existing plant capabilities for accident management, both in-vessel and ex-vessel strategies were identified...

This chapter discusses the development of current state-of-the-art computational methods used in the United States for analysis of core meltdown accidents (Hypothetical Core Disruptive Accidents) in LMFBRs. The emphasis is on the phenomenological basis and numerical methods of the codes SAS, VENUS-II, SIMMER, and REXCO.

Within the framework of a danger analysis which serves as a decision-finding basis for emergency planning, rarely expected severe accidents, which are not covered by design basis accidents, have to be considered at the Rossendorf research site with regard...

Boiling water reactor severe accident sequence studies are being carried out using Browns Ferry Unit 1 as the model plant. Four accident studies were completed, resulting in recommendations for improvements in system design, emergency procedures, and operator training. Computer code improvements were an important by-product.

The publication contains statistical information compiled from reports of 4,648 General Aviation accidents that occurred during the calendar year 1971. Included in the total number of accidents are 51 collisions between aircraft. By coding each aircraft i...

The publication contains statistical information compiled from reports of 4,757 general aviation accidents that occurred during the calendar year 1969. Included in the total number of accidents are 45 collisions between aircraft. By coding each aircraft i...

This paper describes the analysis of the Three Mile Island-2 (TMI-2) standard problem that was performed with MELCOR. The MELCOR computer code is being developed by Sandia National Laboratories for the Nuclear Regulatory Commission for the purpose of analyzing severe accident in nuclear power plants. The primary role of MELCOR is to provide realistic ...

The Severe Accident Analysis Program (SAAP) is a program of experimental and analytical studies aimed at characterizing severe accidents that might occur in the Savannah River Site Production Reactors. The goals of the Severe Accident Analysis Program are: To develop an understanding of severe accidents in SRS ...

A system utilising radiation transport codes has been developed to derive accurate dose distributions in a human body for radiological accidents. A suitable model is quite essential for a numerical analysis. Therefore, two tools were developed to setup a 'problem-dependent' input file, defining a radiation source and an exposed person to simulate the ...

This report identifies the results of a program aimed at developing a computer code for use in the analysis of the behavior of iodine during steam generator tube rupture (SGTR) accidents in pressurized water reactors (PWR's). The program was directed towards the identification of the several processes that play a role in the transport and ...

Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, and fission gas release ...

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss ...

The Transient Reactor Analysis Code (TRAC) is being developed to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in light water reactors. The TRAC-PD2 program provides this analysis capability for pressuriz...

The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged LWR fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and MOX fuels. The following paper describes the derivation, ...

In the present work, a set of codes used for simulations of the radiation fields from ionizing radiation sources inside the containment in an accident is described. A method of evaluating the gamma dose rate from a space and energy distributed source is given. The dose rate is calculated by means of the design point kernel method and using buildup factors. ...

HTGR safety evaluation included studies on fission product release; materials, chemistry, and instrumentation; structural evaluation; and analytical safety evaluation. LMFBR safety evaluation included studies on accident sequences, technical coordination of structural integrity, and SSC code development and validation. LWR safety studies included ...

This circular disseminates, explains, and endorses IMO resolution A.849(20), 'Code for the Investigation of Marine Casualties and Accidents,' which the IMO Assembly adopted on 27 November 1997.

The report describes the two-dimensional, Pressurized Water Reactor (PWR) version of the MELPROG computer code, MELPROG/PWR-MOD1. The purpose of MELPROG is to provide for an accident sequence a description of (1) the state of the reactor core and surround...

A two-dimensional code FRETA-B was developed to analyze the behaviors of LWR fuel rod bundles under accident conditions. It calculates fuel temperatures, oxidation of claddings, plenum gas pressure and flow, and deformations of fuel rods. Two-dimensional ...

This paper summarizes the text and describes the processes followed to develop the first computer-generated film of LASL's Reactor Safety efforts. The 11-1/2 min film with narrative and musical background gives a brief overview of reactor components, of how LASL's Reactor Safety groups develop and verify computer codes to anticipate ...

Calculations of the skyshine gamma-ray dose rates from three spent fuel storage pools under worst case accident conditions have been made using the discrete ordinates code DOT-IV and the Monte Carlo code MORSE and have been compared to those of two previo...

The BWR-LTAS code was developed by the SASA program at Oak Ridge National Laboratory for the detailed study of specific accident sequences at Browns Ferry Unit One: station blackout, small break LOCA outside primary containment, loss of decay heat removal...

The THERMIX-KONVEK code was used to model the steady state and dynamic thermal behavior of the U.S. pebble bed modular HTR concept and trial calculations for accident conditions were performed. Results of these trial calculations are compared with other p...

The Korea Electric Power Research Institute had decided to develop the new safety analysis code system for the Optimized Power Reactor 1000 (OPR1000) in Korea by the fund of the Ministry of Commerce, Industry and Energy. In this paper, some results of the Advanced Power Reactor 1400(APR1400) using the RETRAN code for some reactivity insertion ...

The MAAP Accident Response System (MARS) is a userfriendly computer software developed to provide management and engineering staff with the most needed insights, during actual or simulated accidents, of the current and future conditions of the plant based on current plant data and its trends. To demonstrate the reliability of the MARS ...

The KORSAR best estimate computer code has been developed at NITI since 1996. It is designed to numerically simulate transient and accident conditions in VVER-type reactors /1/. From 1999 and on, the code development activity has been coordinated by the Center for Computer Code Development under Russia's ...

The significant features of SIMMER-II, a disrupted-core analysis code, are described. The code has the capabalities to begin space-time neutronics calculations from nonstationary reactor states, to track the intermixing of fuel of different enrichments, and to model the complicated heat- and mass-transfer processes that occur in the transition phase. ...

Comparisons of results from TRAC-PF1/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters whi...

circulation studies The theoretical skills include � CFD code CFX � Thermal hydraulic system codes RELAP5-coolant interaction: COMETA � Thermal hydraulic system codes, severe accident and reactor core analyses tools: RELAP5 or national thermal hydraulic codes (RELAP5, CATHARE, APROS, ...

A description is presented of the computer code GAPCON THERMAL-2, a light water reactor (LWR) fuel thermal performance prediction code. GAPCON- THERMAL-2, is intended to be used as a calculational tool for reactor fuel steady- state thermal performance and to provide input for accident analyses. Some models used in the ...

Preliminary designs are proposed for extending the SCDAP/RELAP5 code so that it models (a) the oxidation of slumping fuel rod material and cohesive and porous debris and (b) the interaction of PWR control rod materials with the other materials in a reactor core. These extensions have the purpose of improving the code`s calculation of the damage progression ...

Applications of an Eulerian code to predict the response of LMFBR containment and primary piping systems to hypothetical core disruptive accidents (HCDA), and to analyze sodium spillage problems, are described. The computer code is an expanded version of the ICECO code. Sample problems are presented for ...

A calculational procedure for the Station Blackout Severe Accident Sequence at Browns Ferry Unit One has been repeated with plant-specific application to one of the Peach Bottom Units. The only changes required in code input are with regard to the primary continment concrete, the existence of sprays in the secondary containment, and the size of the ...

Response surface techniques have been developed for obtaining probability distributions of the consequences of postulated nuclear reactor accidents. The probabilistic response surface methodology reported includes new knot-point selection schemes and response surface functions, functional transformations of both parameters and consequence variables, smooth synthesis of ...

The radiological analyses of extreme hypothetical accidents were performed almost wholly by computer techniques. Major analytical codes used were RIBD, CACECO, SPRAY, HAA, and COMRADEX. This paper describes the analyses, the modeling techniques, and transition programs that edited data output from one code into a form suitable for ...

The purpose of this study is to provide a parametric framework for characterization of flow and heat transfer regimes and their associated phenomenological uncertainties following severe accidents using a two dimensional, heterogenous, porous media formulation. This approach extends the understanding of buoyancy-induced flow characteristics in the uncovered region of the ...

The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for ...